CP-202400358, Vistra Operations Company, LLC - Proposed Alternative to Use American Society of Mechanical Engineers (ASME) Code Case N-752, Risk-Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 Systems Section

Vistra Operations Company, LLC - Proposed Alternative to Use American Society of Mechanical Engineers (ASME) Code Case N-752, Risk-Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 SystemsSection XI,

ML25091A263
Person / Time
Site:	Beaver Valley, Davis Besse, Perry, Comanche Peak (DPR-066, NPF-073, NPF-087, NPF-089, NPF-003, NPF-058)
Issue date:	04/01/2025
From:	John Lloyd Vistra Operations Company
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
CP-202400358, TXX-24069, L-24-211
Download: ML25091A263 (1)

Text

CP-202400358 TXX-24069 L-24-211 April 1, 2025 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

Beaver Valley Power Station, Unit Nos. 1 and 2 Docket Nos. 50-334 and 50-412 Comanche Peak Nuclear Power Plant, Units 1 and 2 Docket Nos. 50-445 and 50-446 Davis-Besse Nuclear Power Station, Unit No. 1 Docket No. 50-346 Perry Nuclear Power Plant, Unit No. 1 Docket No. 50-440 Jay J. Lloyd Vice President, Nuclear Engineering Vistra Operations Company LLC 6322 North FM 56 Glen Rose, TX 76043 Office: 832.528.1424 10 CFR 50.55a Proposed Alternative to Use American Society of Mechanical Engineers (ASME) Code Case N-752, "Risk-Informed Categorization and Treatment for Repair /Replacement Activities in Class 2 and 3 SystemsSection XI, Division 1" Pursuant to 10 CFR 50.55a(z)(1), Vistra Operations Company LLC (Vistra OpCo) requests Nuclear Regulatory Commission (NRC) authorization to utilize a proposed alternative to the ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components." The requested alternative would be applicable to Beaver Valley Power Station, Unit Nos. 1 (BVPS-1) and 2 (BVPS-2), Comanche Peak Nuclear Power Plant, Unit Nos. 1 (CPNPP-1) and 2 (CPNPP-2), Davis-Besse Nuclear Power Station, Unit No. 1 (DBNPS), and Perry Nuclear Power Plant, Unit No. 1 (PNPP).

Specifically, Vistra OpCo is requesting to use the alternative requirements of ASME Code Case N-752, "Risk-Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 SystemsSection XI, Division 1," for determining the risk-informed categorization and for implementing alternative treatment for repair/replacement activities on moderate and high energy Class 2 and 3 items or components in lieu of certain ASME Code Section XI, paragraph 1W A-1000, 1W A-4000, and 1W A-6000 requirements. Vistra OpCo requests approval on the basis that the proposed alternative provides an acceptable level of quality and safety in accordance with 10 CFR 50.55a(z)(1).

The proposed alternative is provided in the Attachment to this submittal. Vistra OpCo requests NRC approval of the proposed alternative within one year of acceptance.

6555 SIERRA DRIVE IRVING, TEXAS 75039 o 214-812-4600 VISTRACORP.COM

Beaver Valley Power Station, Unit Nos. 1 and 2 Comanche Peak Nuclear Power Plant, Units 1 and 2 Davis-Besse Nuclear Power Station, Unit No. 1 Perry Nuclear Power Plant, Unit No. 1 TXX-24069 L-24-211 Page2 There are no regulatory commitments contained in this submittal. If there are any questions, or if additional information is required, please contact Jack Hicks, Senior Manager, Fleet Licensing, at (254) 897-6725 or jack.hicks@vistracorp.com.

Attachment 10 CFR 50.55a Request Number: VISTRA-ISI-ALT-2024-01 cc:

NRC Region I Administrator NRC Region III Administrator NRC Region IV Administrator Sincerely, Jay J. Lloyd NRC Resident Inspector - Beaver Valley Power Station, Unit Nos. 1 and 2 NRC Resident Inspector - Comanche Peak Nuclear Power Plant, Units 1 and 2 NRC Resident Inspector - Davis-Besse Nuclear Power Station, Unit No. 1 NRC Resident Inspector - Perry Nuclear Power Plant, Unit No. 1 NRR Project Manager - Fleet Director BRP /DEP Site BRP /DEP Representative Utility Radiological Safety Board

Attachment TXX-24069 L-24-211 Page 1 of 24 10 CFR 50.55a Request Number: VISTRA-ISI-AL T-2024-01 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)

-- Alternative Provides Acceptable Level of Quality and Safety --

1. ASME Code Components Affected This request applies to American Society of Mechanical Engineers (ASME) Class 2 and 3 items or components except the following:

Piping within the break exclusion region [> Nominal Pipe Size (NPS) 4 (ON 100)] for high energy piping systems1 as defined by the Owner.

That portion of the Class 2 feedwater system [> NPS 4 (ON 100)] of pressurized water reactors (PWRs) from the steam generator (SG), including the SG, to the outer containment isolation valve.

This request does not apply to Class (CC) Concrete Containment and Class (MC) Metal Containment items.

2. Applicable Code Edition and Addenda

lnservice ASME Plant Inspection Section Interval Interval

.{!§!l XI Code Start End Interval of Record Beaver Valley Power Station, 5th 2013 08/29/2018 08/28/2028 Unit No. 1 {BVPS-1)

Edition Beaver Valley Power Station, 4th 2013 08/29/2018 08/28/2028 Unit No. 2 {BVPS-2)

Edition Comanche Peak Nuclear 4th 2019 08/13/2020 08/12/2030 Power Plant, Unit 1 (CPNPP-1)

Edition Comanche Peak Nuclear 4th 2019 08/03/2023 08/02/2033 Power Plant, Unit 2 {CPNPP-2)

Edition Davis-Besse Nuclear Power 5th 2017 06/08/2023 09/20/2032 Station, Unit No. 1 {DBNPS)

Edition Perry Nuclear Power Plant, 4th 2013 05/18/2019 05/17/2029 Unit No. 1 (PNPP)

Edition 1 NUREG-0800, Section 3.6.2 provides a method for defining this scope of piping.

Attachment TXX-24069 L-24-211 Page 2 of 24

3. Applicable Code Requirements

ASME Code,Section XI, Subsection IWA provides the requirements for repair/replacement activities including the following:

IWA-1320 specifies group classification criteria for applying the rules of ASME Section XI to various Code Classes of components. For example, the rules in IWC apply to items classified as ASME Class 2 and the rules in IWD apply to items classified as ASME Class 3.

IWA-1400(g)2 requires Owners to possess or obtain an arrangement with an Authorized Inspection Agency (AIA).

IWA-1400(k)2 requires Owners to perform repair/replacement activities in accordance with written programs and plans.

IWA-1400(0)2 requires Owners to maintain documentation of a Quality Assurance Program in accordance with 10 CFR 50 or ASME NQA-1, Part I.

IWA-4000 specifies requirements for performing ASME Section XI repair/replacement activities on pressure-retaining items or their supports.

IWA-6211 (d)2, IWA-6211 (e)2, and IWA-62202 specify Owner reporting responsibilities such as preparing Form NIS-2, Owner's Report for Repair/Replacement Activity.

IWA-6211 (f)2 and IWA-6212 specify Repair/Replacement Organization documentation and certification requirements.

IWA-6350 specifies that the following ASME Section XI repair/replacement activity records must be retained by the Owner: evaluations required by IWA-4160 and IWA-4311, Repair/Replacement Programs and Plans, reconciliation documentation, and NIS-2 Forms.

4. Reason for Request

Vistra Operations Company LLC (Vistra OpCo) currently performs repair/replacement activities in accordance with deterministic Repair/Replacement Programs based on the following: CPNPP-1 and CPNPP-2, ASME Section XI 2019 Edition; DBNPS, ASME Section XI 2017 Edition; and PNPP, BVPS-1, and BVPS-2, ASME Section XI 2013 Edition.

Repair/Replacement Program requirements apply to procurement, design, fabrication, 2 Code Case N-752 is based on the 2017 Edition of ASME Section XI while Vistra OpCo's Code of Record for CPNPP-1 and CPNPP-2 is the 2019 Edition, DBNPS is the 2017 Edition, and PNPP, BVPS-1, and BVPS-2 is the 2013 Edition of ASME Section XI, Division 1. Paragraph references in Section 3, above, are applicable to the 2013, 2017, and 2019 Editions of ASME Section XI with one exception: IWA-6220 in the 2017 Edition does not exist in or apply to the 2013 Edition.

Attachment TXX-24069 L-24-211 Page 3 of 24 installation, examination, and pressure testing of items within the scope of ASME Section XI. Repair/replacement activities include welding, brazing, defect removal, metal removal using thermal processes, rerating, and removing, adding, or modifying pressure retaining items or supports. Repair/replacement activities are performed in accordance with the plant's 10 CFR 50, Appendix B, Quality Assurance (QA) Program and the ASME Section XI Code. In applying a deterministic approach to repair/replacement activities, a safety class (e.g., ASME Class 2 or 3) is assigned to every component within a system based on system function; the same treatment requirements are then applied to every component within the system without considering the risk associated with the probability that a specific item or component may or may not be functional at a time when needed.

Alternatively, a probabilistic approach to regulation enhances and extends the traditional deterministic approach by allowing consideration of a broader set of potential challenges to safety, providing a logical means for prioritizing these challenges based on safety significance, and allowing consideration of a broader set of resources to defend against these challenges. In contrast to the deterministic approach, probabilistic risk assessment (PRA) addresses credible initiating events by assessing the event frequency. Mitigating system reliability is then assessed, including the potential for common cause failures. The probabilistic approach to regulation is an extension and enhancement of traditional regulation by considering risk in a comprehensive manner. In 2004, the Nuclear Regulatory Commission (NRC) adopted a new Section 50.69 of 10 CFR relating to risk-informed categorization and treatment of structures, systems, and components (SSCs) for nuclear power plants (Reference 8.1). This new section permits power reactor licensees to implement an alternative regulatory framework with respect to "special treatment" (treatment beyond normal industrial practices) of low safety significant (LSS) SSCs. In May 2006, the NRC staff issued Regulatory Guide (RG) 1.201, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance, For Trial Use," Revision 1 (Reference 8.2). RG 1.201 endorses a categorization method, with conditions, for categorizing active SSCs described in Nuclear Energy Institute (NEI) 00-04, "10 CFR 50.69 SSC Categorization Guideline."

Vistra OpCo is not requesting NRC approval to implement 10 CFR 50.69 in this relief request. Instead, Vistra OpCo is proposing to implement the risk-informed categorization and treatment requirements of ASME Code Case N-752 when performing repair/replacement activities on Class 2 and 3 pressure-retaining items or their associated supports. Code Case N-752, which was approved by the ASME in July 2019, employs a comprehensive categorization process requiring input from both a PRA model and deterministic insights. This approach will enable evaluation, categorization, and implementation of alternative treatments for resolution of emergent issues in segments of piping having low safety significance. Use of Code Case N-752 will also allow Vistra OpCo to identify and more clearly focus engineering, maintenance, and operations resources on critical components with high safety-significance, thus, enabling Vistra OpCo to make more informed decisions and increase the safety of the plant.

Attachment TXX-24069 L-24-211 Page 4 of 24

5. Proposed Alternative and Basis for Use

Pursuant to 10 CFR 50.55a(z)(1), Vistra OpCo proposes to implement ASME Code Case N-752 with no exceptions or deviations. ASME Code Case N-752 would be used as an alternative to the ASME Code requirements specified in Section 3 of this request. Code Case N-752 provides a process for determining the risk-informed categorization and treatment requirements for Class 2 and 3 pressure-retaining items or the associated supports as delineated in Section 1 of this request. This requested implementation includes the categorization of passive SCCs (e.g., piping) and implementation of alternative special treatment activities limited to the repair/replacement activities for Class 2 and 3 pressure-retaining items or their associated supports. For components that have both active and passive functions, only the passive function will be categorized. The alternative treatments associated with ASME Code Case N-752 will not be applied to the parts/components associated with the active function. Code Case N-752 may be applied on a system basis or on individual items within selected systems. Code Case N-752 does not apply to Class 1 items.

The use of this proposed alternative is requested on the basis that requirements in Code Case N-752 will provide an acceptable level of quality and safety.

5.1 Overview of Code Case N-752 Code Case N-752 provides for risk-informed categorization and treatment requirements for performing repair/replacement activities on Class 2 and 3 pressure-retaining items or their associated supports. Code Case N-752 is not applicable to the following:

Class CC and MC items.

Piping within the break exclusion region [> NPS 4 (ON 100)] for high energy piping systems as defined by the Owner.

That portion of the Class 2 feedwater system [> NPS 4 (ON 100)] of PWRs from the SG, including the SG, to the outer containment isolation valve.

Code Case N-752 categorization methodology relies on the conditional core damage and large early release probabilities associated with postulated ruptures. Safety significance is generally measured by the frequency and the consequence of the event. However, the risk-informed process categorizes components solely based on consequence, which measures the safety significance of the component given that it ruptures (component failure is assumed with a probability of 1.0). This approach is conservative compared to including the rupture frequency in the categorization as this approach will not allow the categorization of SSCs to be affected by any changes in frequency due to changes in treatment. It additionally applies deterministic considerations (e.g., defense in depth, safety margins) in determining safety significance. Additional detail is provided in Section 5.2 of this request.

Attachment TXX-24069 L-24-211 Page 5 of 24 The risk-informed process categorizes components as either high safety-significant (HSS) or LSS. HSS components must continue to meet ASME Section XI rules for repair/replacement activities. LSS components can be exempted from ASME Section XI repair/replacement requirements and can be repaired/replaced in accordance with treatment requirements established by the Owner. The treatment requirements must provide reasonable confidence that each LSS item remains capable of performing its safety-related functions under design basis conditions. Component supports, if categorized, are assigned the same safety significance, HSS or LSS, as the highest passively ranked segment within the bounds of the associated analytical pipe stress model. The categorization and treatment requirements of Code Case N-752 are consistent with those in 10 CFR 50.69.

Code Case N-752 is based on Arkansas Nuclear One (ANO), Unit 2 (ANO-2) relief request ANO2-R&R-004, Revision 1, dated April 17, 2007 (Reference 8.3), as supplemented by Entergy. The NRC approved relief request ANO2-R&R-004, Revision 1, in a safety evaluation dated April 22, 2009 (Reference 8.4). The ANO-2 relief request was developed to serve as an industry pilot for implementing a risk-informed repair/replacement process that included a risk-informed categorization process and treatment requirements.

5.2 Basis for Use The information below is provided as a basis or justification for Vistra OpCo's proposed alternative to implement the risk-informed categorization and treatment requirements of Code Case N-752 on Class 2 and 3 pressure-retaining items or the associated supports as delineated in Section 1 of this request.

A.

Application to Individual Items Within a System The risk-informed methodology of Code Case N-752 may be applied on a system basis or on individual items within selected systems. Paragraph -1100 of Code Case N-752 states: "This Case may be applied on a system basis, including all pressure-retaining items and their associated supports, or on individual items categorized as low-safety-significant (LSS) within the selected systems." While this is the case, the risk-informed methodology is, in actuality, applied to the pressure boundary function of the individual components within the system. The risk-informed methodology contained in Code Case N-752 requires that the component's pressure boundary function be assumed to fail with a probability of 1.0, and all impacts caused by the loss of the pressure boundary function be identified. This would include identifying impacts of the pressure boundary failure on the component under evaluation, identifying impacts of the pressure boundary failure of the component on the system in which the component resides, as well as identifying impacts of the pressure boundary failure of the component on any other plant SSC. This includes direct effects (e.g., loss of the flow path) of the component failure and indirect effects of the component failure (e.g., flooding,

Attachment TXX-24069 L-24-211 Page 6 of 24 spray, pipe whip, loss of inventory). This comprehensive assessment of total plant impact caused by a postulated individual component failure is then used to determine the final consequence ranking. As such, the final consequence rank of the individual component would be the same regardless of whether the entire system or only the individual component is subject to the risk-informed methodology.

B.

Categorization Process The categorization process of Code Case N-752 is delineated in Appendix I of the Code Case. This categorization process is technically identical to the process approved by the NRC under Relief Request ANO2-R&R-004, Revision 1 (Reference 8.3), which, in turn, is based on founding principles in Electric Power Research Institute (EPRI) Report TR-112657, Revision 8-A, "Revised Risk-Informed lnservice Inspection Evaluation Procedure," (Reference 8.5) and the categorization process of Code Case N-660, but with improvements and lessons learned from trial applications. ASME Code Case N-752 also provides a definition of piping segment in -9000 Glossary, which Vistra OpCo will utilize to define a pipe segment for categorization.

The Code Case N-752 risk-informed categorization evaluation is performed by an Owner-defined team that includes experts with expertise in PRA, plant operations, system design, and safety or accident analysis. The risk-informed categorization process is based on the conditional consequence of failure, given that a postulated failure has occurred. A consequence category for each piping segment or component is determined via a failure modes and effects analysis (FMEA) and impact group assessment. The FMEA considers pressure boundary failure size, isolability of the break, indirect effects, initiating events, system impact or recovery, and system redundancy. The results of the FMEA for each system, or portion thereof, are partitioned into core damage impact groups based on postulated piping failures that cause an (1) initiating event, (2) disable a system/train/loop without causing an initiating event, or (3) cause an initiating event and disable a system/train/loop. Failures are also evaluated for their importance relative to containment performance. In addition, the consequence rank is reviewed and adjusted to reflect the pressure boundary failure's impact on plant operation during shutdown and on the mitigation of external events.

Credit may be taken for plant features and operator actions to the extent these would not be adversely affected by failure of the piping segment or component under consideration.

Consequence evaluation results are ranked as High, Medium, Low, or None (no change to base case). Piping segments/components ranked as High by the consequence evaluation process are considered HSS and require no further review. Piping segments/components ranked as Medium, Low, or None by the consequence evaluation shall be determined to be HSS or LSS by evaluating the additional categorization considerations or conditions outlined in paragraph

Attachment TXX-24069 L-24-211 Page 7 of 24 l-3.4.2(b) of Code Case N-752. If any of these conditions are not met, then HSS shall be assigned. If all conditions are met, then LSS may be assigned. Finally, if LSS is assigned, the categorization process shall verify that there are sufficient margins to account for uncertainty in the engineering analysis and supporting data. If sufficient margin exists, then LSS should be assigned. If sufficient margin does not exist, then HSS shall be assigned.

C.

PRA Technical Adequacy Appendix I, Section 1-3.2 of Code Case N-752 requires that the plant-specific PRA shall be assessed to confirm it is applicable to the safety significant categorization of ASME Code Case N-752 including verification of assumptions on equipment reliability for equipment not within the scope of the ASME Code Case.

BVPS-1 and BVPS-2 PRA Technical Adequacy The Beaver Valley PRA models were previously used to support the Risk-I nformed In-Service Inspection (RI-ISi) program, initially approved by the NRC in April 2004 (Reference 8.6). Implementation of this program had similar requirements, as the methodology for Risk-Informed Repair and Replacement has its foundation in EPRI Report TR-112657 (Reference 8.5).

The Beaver Valley PRA models were also more recently used in the approval of Risk-Informed Surveillance Frequency Control Program (SFCP) (Reference 8. 7) and 10 CFR 50.48(c) National Fire Protection Association (NFPA) 805 (Reference 8.8). All peer review Facts & Observations (F&Os) prior to 2014 and their resolutions were submitted in detail and were ultimately found satisfactory for those reviews. An F&O Closure Independent Assessment has subsequently been performed to formally close most of those F&Os.

The BVPS-1 and BVPS-2 PRA models have been the subject of several assessments to establish the technical adequacy of the PRA. These assessments regarding internal events and internal flooding are identified and discussed in the paragraphs below. Peer reviews of the BVPS-1 and BVPS-2 Seismic PRA and Fire PRA models were also performed but are not directly relevant to the scope of this application and are not included here.

2002 - An independent PRA peer review of the Beaver Valley PRA models was conducted under the auspices of the Westinghouse Owners Group (WOG) in July 2002, following the NEI 00-02 Industry PRA Peer Review process. This peer review included an assessment of the PRA model maintenance and update process.

Attachment TXX-24069 L-24-211 Page 8 of 24 2007 - Following the BVPS-1 PRA model revision in 2006 and the BVPS-2 PRA model revision in 2007, an independent assessment of the Beaver Valley PRA models against the ASME/ANS RA-Sb-2005 PRA Standard was performed by Westinghouse personnel using RG 1.200, Revision 1. This review assessed the work performed to close F&Os from the original peer review and wrote new F&Os for any previously identified issues, which remained unresolved at the time, as well as for any newly identified issues against compliance with Capability Category II or higher of the ASME/ANS Standard. Thus, the results of this independent assessment effectively replace the results from the original 2002 Peer Review.

2007 -As part of the resolution to several F&Os from the 2002 PRA peer review, a change in the Human Reliability Analysis (HRA) methodology was incorporated into the 2006 BVPS-1 PRA model revision and the 2007 BVPS-2 PRA model revision, so a focused scope peer review of the HRA Technical Elements against the ASME/ANS RA-Sb-2005 PRA Standard was performed using RG 1.200, Revision 1, in accordance with the NEI 05-04 process to address the upgrade. The results of this peer review replace the results from the HR high level technical requirements from the previous 2007 independent assessment.

2011 - Due to an upgrade of the Internal Flooding model following the BVPS-1 and BVPS-2 PRA model revisions in 2010, a focused scope peer review of the Internal Flooding PRA Technical Elements was performed against the applicable requirements of Part 3 of the ASME/ANS PRA Standard RA-Sa-2009 (along with the NRC clarifications provided in RG 1.200, Revision 2). The results of this peer review replace the results from the Internal Flooding high level technical requirement from the previous 2007 independent assessment.

2017-An F&O Closure Independent Assessment of the BVPS-1 and BVPS-2 PRA models was performed in accordance with Appendix X of NEI 05-04/07-12/12-13. This resulted in the closure of all previous Internal Events and Internal Flooding F&Os against the Beaver Valley PRA models except for one Internal Events F&O and one Internal Flooding F&O. These two F&Os, which remain open, are discussed relative to this application in Table 1.

The current BVPS-1 Revision 8 (2023) PRA model (PRA-BV1-AL-R08) and the current BVPS-2 Revision 8 (2022) PRA model (PRA-BV2-AL-R08) have each resolved all the F&Os remaining from the 2017 BVPS PRA F&O Closure Independent Assessment. These PRA models are compliant with RG 1.200, Revision 2, for the scope of this application and meet Capability Category II or above in the ASME PRA Standard (RA-Sa-2009).

Attachment TXX-24069 L-24-211 Page 9 of 24 The remaining Internal Events and Internal Flooding F&Os against the Beaver Valley PRA models are show in Table 1 along with their resolutions. These issues have been resolved to achieve Capability Category II or higher and thus have no impact on any evaluation pertaining to Code Case N-752. The open Internal Events F&O and its resolution has been previously reviewed by NRC staff in the applications for NFPA 805 as referenced above and was determined to have been acceptably addressed to satisfy Capability Category II or higher as necessary for that application. The open Internal Flooding F&O only remains open because of its tie to the open Internal Events F&O.

Therefore, the Beaver Valley PRA models are of adequate technical capability to support Code Case N-752.

Attachment TXX-24069 L-24-211 Page 10 of 24 F&O HR-PR-001 Table 1: Open F&Os in Beaver Valley Revision 8 Internal Events and Internal Flooding PRA Models Related Related F&OSummary BVPS Resolution Supporting SRs:

Summary Requirements ASME/ANS (SRs):

RA-Sa-2009 ASME/ANS RA-Sb-2005 HR-D5 HR-D5 BVPS does not have a SRs HR-D5, HR-H3, HR-G7 HR-G7 written process for HR-I1, and HR-I2 were all HR-H3 HR-H3 evaluating dependencies separately re-evaluated as HR-I1 HR-I1 between multiple HEPs MET in the 2017 F&O HR-I2 HR-I2 occurring in a single Closure Independent accident and does not Assessment. SR HR-G7 provide a summary of and F&O HR-PR-001 HEPs that were explicitly were not closed as part of evaluated for dependencies that assessment.

and the associated levels of dependencies and joint Section 2.2 of the HRA HEPs. The BVPS HRA Notebook was added to notebooks do not have a document the single summary table of the methodology and pre-initiator human actions evaluation of the pre-and the documentation of initiator HEPs. A summary the evaluation of pre-of the EPRI HRA initiator human actions in Calculator results for pre-the system notebooks, initiators can now be which make it difficult to found in Table 3-6, which identify which actions were supplements the detailed actually evaluated.

calculations documented in Appendix E.

Section 2.3 of the HRA notebook was added to discuss the dependency analysis and refers to the separate PRA Assessments PRA-BV1-19-004-R00 I PRA-BV2-20-002-R00. The process employing the EPRI HRA calculator used to complete the dependency analysis evaluation is detailed for each unit in these Assessments. This process was also discussed as part of the NRC NFPA 805 LAR Audit and RAI cycle and was found acceptable for NFPA 805.

Attachment TXX-24069 L-24-211 Page 11 of 24 F&O IFQU-AS-01 Table 1: Open F&Os in Beaver Valley Revision 8 Internal Events and Internal Flooding PRA Models Related Related F&OSummary BVPS Resolution Supporting SRs:

Summary Requirements ASME/ANS (SRs):

RA-Sa-2009 ASME/ANS RA-Sb-2005 IF-ES IFQU-A5 It appears that no inter-As a resolution to this HEP dependency analysis IFPRA Peer Review (between flood and non-finding, Section 10.4.6 flood HEPs) was (Dependencies between performed. Dependency Human Interactions) was between H EPs can revised to state that an significantly increase the HRA dependency analysis probabilities of was performed and that a combinations of HEPs.

discussion on the "HFE However, Section 10.4 of Dependencies in Internal the internal flooding PRA Flooding PRAAccident reports states Sequences" is provided in "Dependencies between the HRA Notebook the flood mitigation human Section 2.3.

actions and the non-flood Section 10.4.6 of the human actions modeled in the remaining part of the Internal Flooding PRA model were judged to Notebook was expanded be minimal due to the to reiterate Section 10.4.3 significant difference in the (Screening and Detailed nature of the actions (e.g.,

Analysis) discussion on flood mitigation actions the multiplier factor require field investigation applied to HEPs included by the auxiliary operators, in the Internal Events PRA based on such factors as etc.) and separation in time, the location of the action, etc., and as such no additional dependency the timing of the action, treatment was considered and stress, etc., and to needed." An evaluation of include a discussion of the the HEP combinations Riskman modeling should be documented to analysis approach in demonstrate this which human actions conclusion.

included are evaluated conditionally based on the success or failure status of the preceding human action(s). As such, dependencies among the human failure events in the Internal Events model (i.e., non-flood human actions) were fully accounted.

Resolution of this F&O is tied to the resolution of F&O HR-PR-001.

Attachment TXX-24069 L-24-211 Page 12 of 24 CPNPP-1 and CPNPP-2 PRA Technical Adequacy The CPNPP-1 and CPNPP-2 PRA model was previously used in the NRC approval of the Risk-Informed Completion Time (TSTF-505, Revision 2) application (Reference 8.9) and the approval of the 10 CFR 50.69 (Reference 8.10). The Comanche Peak PRA model consists of the Internal Events model, the Internal Flood model, and the Internal Fire model. Each model has been peer-reviewed, and the relevant F&Os have been formally closed through the NEI 05-04 Appendix X process.

The Comanche Peak PRA model is sufficiently robust and suitable for use in risk-informed applications such as for regulatory decision making. The peer reviews that have been conducted and the resolution of findings from those reviews demonstrate that the internal events, internal flooding, and fire models of the PRA have been performed in a technically correct manner. The assumptions and approximations used in development of the PRA have also been reviewed and are appropriate for this application.

Comanche Peak has established PRA Configuration Control procedures to provide assurance the PRA models continue to reflect the as-built, as-operated and maintained plant. These reviews include quarterly reviews of plant changes (design and procedures) as part of the Mitigating System Performance Index process, the 36-month reviews as part of the 50.69 process, and the 48-month update frequency required by the Risk-Informed Completion Time programs.

The above discussion demonstrates that the CPNPP-1 and CPNPP-2 PRA models are technically acceptable to support the ASME Code Case N-752 application.

DBNPS PRA Technical Adequacy The DBNPS PRA models were previously used to support the approval of Risk-Informed SFCP (Reference 8.11) and 10 CFR 50.48(c) NFPA 805 (Reference 8.12).

Specific supporting requirements (SRs) noted in the following discussion are identified by their ASME/ANS RA-Sa-2009 designation, followed by the ASME/ANS RA-Sa-2005 designation in parentheses.

An independent assessment and closeout review were performed in October 2017 to address the DBNPS disposition to all F&Os from all previous peer reviews. The review was conducted consistent with NEI 05-04/07-12/12-13 Appendix X, "Fact & Observation (F&O) Close-Out with Independent Assessment." Each F&O documented closure was reviewed to determine if the F&O had been adequately addressed and could therefore be closed out using the appropriate parts of the AS ME/ANS RA-Sa-2009 PRA standard. The relevant

Attachment TXX-24069 L-24-211 Page 13 of 24 SRs were also reassessed for cases where the peer review identified the SR as not meeting Capability Category II. It was determined the DBNPS PRA Internal Events and Internal Flooding models meet Capability Category II for all SRs except for one open finding, SY-811, which is open for SY-810 (SY-811) to meet Capability Category II. However, SY-810 (SY-811) was determined to meet Capability Category I. In addressing F&O SY-811 in the model, it was determined that the Class 1-E 4160V [volt] bus low voltage signal and safety features actuation system (SFAS) sequencer permissive actuation logic were the only additional modeling and documentation changes needed to address the finding. The changes have been made in the model, but an independent assessment has not been performed to formally close the F&O and formally assess SY-810 (SY-811) as Capability Category II. These modeling changes are considered model maintenance per the definition in the ASME/ANS PRA Standard. SY-810 (SY-811) is required to be met at Capability Category I to support Code Case N-752. Thus, the DBNPS Internal Events and Internal Flooding PRA is of adequate technical capability to support Code Case N-752.

PNPP PRA Technical Adequacy The PNPP PRA model was previously used in the approval of the Risk-Informed SFCP (Reference 8.13).

The PNPP PRA model has undergone numerous peer reviews and self-assessments. A formal peer review of the model was performed initially in 1997 under the Boiling Water Reactor (BWR) Owner's Group and only mentioned here for the sake of completeness. This peer review was superseded, in accordance with NEI 05-04 (Reference 8.14)Section X.1.1, in 2007 by a peer review to the ASME RA-Sb-2005 PRA Standard (Reference 8.15) and RG 1.200, Revision 1 (Reference 8.16). The 2007 review met all the requirements for the conduct of a peer review as described in NEI 05-04 (Reference 8.17). The Large Early Release and Internal Flooding portions later had separate focused scope peer reviews against the relevant portions of the ASME/ANS RA-Sa-2009 PRA Standard (Reference 8.18) and RG 1.200, Revision 2 (Reference 8.19). The Large Early Release portion of the model was peer reviewed in November 2011, and the Internal Flooding portion of the model was peer reviewed in July 2012.

An additional focused scope peer review was performed in July 2015 covering specific Supporting Requirements (SRs) pertinent to the recovery of offsite power, assessed against ASME/ANS PRA Standard RA-Sb-2013 (Reference 8.20), using the NEI 05-04 process, and the NRC clarifications provided in RG 1.200, Revision 2. While this review was done to ASME/ANS PRA Standard RA-Sb-2013, the changes between this Standard and ASME/ANS RA-Sa-2009 for the Part 2 requirements under the scope of this peer review are minimal.

A Gap Assessment between RG 1.200, Revision 2, and ASME/ANS RA-Sa-2009, Addenda to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant

Attachment TXX-24069 L-24-211 Page 14 of 24 Applications, from the previous assessment performed to RG 1.200, Revision 1, and the ASME/ANS RA-Sb-2005, Addenda to ASME RA-S-2002 Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, was performed in 2015 in support of the SFCP request (Reference 8.13).

In October 2017, an Independent Assessment was performed to conduct an F&O Closure Workshop to close out open Findings identified in the previous Peer Reviews. This F&O review and closure followed the guidance and expectations provided in NEI 05-04, NEI 07-12, and NEI 12-13 Appendix X (Reference 8.14),

as well as the expectations of the NRC (Reference 8.21). All finding level F&Os for the internal events and internal flooding model were closed using the NEI Appendix X process. During the performance of the F&O Closure Workshop, the resolution of some Findings were determined by the Independent Assessment team to constitute PRA Upgrades rather than PRA Updates. Therefore, a Focused Scope Peer Review was performed in conjunction with the F&O Closure Workshop to assess portions of the PRA model to the relevant SRs. It was determined the PNPP PRA Internal Events and Internal Flooding models meet Capability Category II for all SRs, with no new Findings identified. The results of this review have been documented and are available for NRC audit.

The above demonstrates that the PNPP Internal Events and Internal Flooding PRA is of adequate technical capability to support Code Case N-752.

D.

Feedback and Process Adjustment Vistra OpCo shall review changes to the plant, operational practices, applicable plant and industry operational experience, and, as appropriate, update the PRA and categorization and treatment processes. Vistra OpCo shall perform this review in a timely manner but no longer than once every two refueling outages.

This approach is consistent with the feedback and adjustment process of 10 CFR 50.69(e).

E.

Treatment Requirements for LSS Items Code Case N-752 exempts LSS items, which have been categorized as LSS in accordance with the code case, from having to comply with the repair/replacement requirements of ASME Section XI. Exempted ASME Code requirements for LSS items are outlined in Section 3, above. In lieu of these requirements, Code Case N-752, Paragraph -1420 requires the Owner to define alternative treatment requirements that confirm with reasonable confidence that each LSS item remains capable of performing its safety-related functions under design basis conditions. These Owner treatment requirements must address or include all of the provisions stipulated in Paragraphs -1420(a) through U) of the code case. This approach to treatment is consistent with Risk-Informed Safety Class (RISC)-3 treatment requirements specified in 10 CFR 50.69(d)(2).

Attachment TXX-24069 L-24-211 Page 15 of 24 To comply with the above, Vistra OpCo intends to develop new and/or revise existing procedures and documents to define treatment requirements for performing repair/replacement activities on LSS items in accordance with Code Case N-752. Vistra OpCo defined treatment requirements are to address design control, procurement, installation, configuration control, and corrective action.

Vistra OpCo procedures and documents are also to include provisions that address/implement the following requirements:

1.

Administrative controls for performing these repair/replacement activities.

2.

The fracture toughness requirements of the original Construction Code and Owner's Requirements shall be met.

3.

Changes in configuration, design, materials, fabrication, examination, and pressure-testing requirements used in the repair/replacement activity shall be evaluated, as applicable, to ensure the structural integrity and leak tightness of the system are sufficient to support the design bases functional requirements of the system.

4.

Items used for repair/replacement activities shall meet the Owner's Requirements or revised Owner's Requirements as permitted by the licensing basis.

5.

Items used for repair/replacement activities shall meet the Construction Code to which the original item was constructed. Alternatively, items used for repair/replacement activities shall meet the technical requirements of a nationally recognized code, standard, or specification applicable to that item as permitted by the licensing basis.

6.

The repair methods of nationally recognized post-construction codes and standards (e.g., PCC-2, APl-653) applicable to the item may be used.

7.

Performance of repair/replacement activities, and associated non-destructive examination (NOE), shall be in accordance with the Owner's Requirements and, as applicable, the Construction Code, or post-construction code or standard, selected for the repair/replacement activity.

Alternative examination methods may be used as approved by the Owner.

NOE personnel may be qualified in accordance with IWA-2300 in lieu of the Construction Code.

8.

Pressure testing of the repair/replacement activity shall be performed in accordance with the requirements of the Construction Code selected for the repair/replacement activity or shall be established by the Owner.

9.

Baseline examination (e.g., preservice examination) of the items affected by the repair/replacement activity, if required, shall be performed in accordance

Attachment TXX-24069 L-24-211 Page 16 of 24 with requirements of the applicable program(s) specifying periodic inspection of items. See paragraph 5.2.E.11, below, for additional details.

10. Implementation of Code Case N-752 does not negate or affect Vistra OpCo commitments to regulatory and enforcement authorities having jurisdiction at BVPS-1, BVPS-2, CPNPP-1, CPNPP-2, DBNPS, and PNPP.

11. Periodic ISi and inservice testing (1ST) of LSS items at BVPS-1, BVPS-2, CPNPP-1, CPNPP-2, DBNPS, and PNPP will continue to be performed as follows:

ISi of LSS pressure-retaining items or their associated supports will be performed in accordance with the site's ISi program implemented in accordance with 10 CFR 50.55a.

1ST of pumps and valves that have been classified as LSS will be performed in accordance with the site's 1ST program implemented in accordance with 10 CFR 50.55a.

1ST of snubbers that have been classified as LSS will be performed in accordance with the site's Snubber Testing program implemented in accordance with 10 CFR 50.55a.

Inspections of LSS items performed under other plant programs, such as the Flow Accelerated Corrosion program, will continue to be performed under those programs for the site.

12. Conditions that would prevent an LSS item from performing its safety-related function(s) under design basis conditions will be corrected in a timely manner. For significant conditions adverse to quality, measures will be taken to provide reasonable confidence that the cause of the condition is determined, and corrective action taken to preclude repetition. Corrective action of adverse conditions associated with LSS items will be identified and addressed in accordance with the existing Vistra OpCo corrective action program that is applicable for each site. Finally, this approach to corrective action of LSS items is consistent with the NRC position on corrective action of RISC-3 SSCs as specified in 10 CFR 50.69(d)(2)(ii).

Technical Specifications required by 10 CFR 50.36 will continue to be complied with in every respect. Specifically, in accordance with Limiting Condition for Operation (LCO) 3.0.1, LCOs will be met during the Modes or other specified conditions in the Applicability. This includes LCOs containing SSCs that have been categorized as LSS. If an LCO is not met during the Modes or other specified conditions in the Applicability, the Required Actions of the associated Conditions will be met in accordance with LCO 3.0.2. The Required Actions would be taken within the associated

Attachment TXX-24069 L-24-211 Page 17 of 24 Completion Times. Required Actions must be completed prior to the expiration of the specified Completion Time.

13. As permitted by Code Case N-752, Vistra OpCo intends to implement the exemption on IWA-1400(g) and IWA-4000 applicable to utilization of an Authorized Insurance Agency (AIA) and Authorized Nuclear lnservice Inspector (ANII) when performing repair/replacement activities on LSS items. In lieu of ANII inspection services, Vistra OpCo believes that its proposed treatment requirements, as described herein, provide reasonable confidence that LSS systems and items remain capable of performing their safety-related functions when repair/replacement activities are performed without the inspection services of an ANII. The exemption of ANII services is not unique to Code Case N-752. Utilization of ANII inspection services is already exempt by ASME Section XI for certain items and activities such as small items (IWA-4131) and rotation of items for testing or preventative maintenance (IWA-4132). Finally, exemption of AIA/ANII services for this code case application is consistent with the NRC's position on risk-informed programs as specified in 10 CFR 50.69(b)(1)(v).

14. Code Case N-752 does not exempt Owners from ASME Section XI subparagraph IWA-1400(0) if compliance with 10 CFR 50, Appendix B, or NQA-1 is required at the Owner's facility, which is the case for the Vistra OpCo nuclear facilities. The QA Program requirements currently applicable to CPNPP-1 and CPNPP-2 are addressed in Chapter 17 of the Comanche Peak Nuclear Power Plant Final Safety Analysis Report (FSAR), which also requires the establishment of the Comanche Peak Nuclear Power Plant Quality Assurance Manual (QAM). The QA Program requirements currently applicable for BVPS-1, BVPS-2, DBNPS, and PNPP are addressed in one combined QA Program Manual (QAPM). Changes to the Owner's QA Program are subject to the requirements of 10 CFR 50.54(a). If the changes do not reduce commitments in the QA program, then the change may be performed in accordance with the provisions of 10 CFR 50.54(a)(3). Based on these criteria, Vistra OpCo intends to amend the QAM and QAPM in accordance with 10 CFR 50.54(a)(3) to address the performance of risk-informed repair/replacement activities in accordance with Code Case N-752.

Specifically, the QAM and QAPM will be revised to include the following:

Based on NRC authorization to use the alternative repair/replacement categorization and treatment requirements of ASME Code Case N-752 in lieu of the corresponding sections of ASME Section XI, as referenced in 10 CFR 50.55a Codes and Standards, treatment of safety-related structures, systems and components (SSCs) identified as low safety significant (LSS) Class 2 and 3 SSCs in accordance with ASME Code Case N-752 is not required to meet the requirements of this document.

Attachment TXX-24069 L-24-211 Page 18 of 24 Instead, treatment of these LSS SSCs is performed in accordance with existing QA Program procedures and processes which include supplemental controls to ensure the capability and reliability of the SSCs design basis function.

The basis for the QA Program change is established in the precedent identified in Section 7.4 of this alternative request and in accordance with 10 CFR 50.54(a)(3)(ii), which establishes that a quality assurance alternative or exception approved by an NRC safety evaluation is not considered a reduction in QA Program commitments provided the bases of the NRC approval are applicable to the licensee's facility. Consistent with the precedent in Section 7.4, under the amended QA Programs, Vistra OpCo will define alternative treatment requirements that confirm with reasonable confidence that each Class 2 and 3 LSS SSC will remain capable of performing its safety-related function under design basis conditions. In doing so, Vistra OpCo will use current QA Program processes and procedures with additional controls for the treatment of Class 2 and 3 LSS components to reasonably assure continued capability and reliability of the design basis function(s). This includes confirming, with reasonable confidence, that changes to the configuration, design, material, fabrication, examination, and testing requirements used to support repair/replacement activities on Class 2 and Class 3 LSS SSCs are performed in accordance with Vistra OpCo's design change process and addressing in Vistra OpCo's corrective action program any condition that may prevent a LSS SSC from performing its design basis function. For the procurement of Class 2 and 3 LSS components as non-safety related for the repair/replacement activities in accordance with ASME's Code Case N-752, supplemental procurement requirements will be specified, and additional controls will be implemented as appropriate to provide reasonable assurance that Class 2 and 3 LSS SSCs will remain capable of performing their safety-related function under design basis conditions. Such controls include conducting receipt inspections using qualified inspection personnel consistent with Vistra OpCo's procurement requirements and prohibiting suppliers of Class 2 and 3 SSCs and subparts from making design changes to the procurement order without prior Vistra OpCo approval. Using these existing QA Program processes and alternative treatment requirements, Vistra OpCo believes that the implementation of ASME Code Case N-752 will provide reasonable confidence that each Class 2 and 3 LSS SSC remains capable of performing its design-basis function.

15. As permitted by Code Case N-752, Vistra OpCo intends to implement the exemptions on IWA-1400(k) and IWA-4000 applicable to repair/replacement programs and plans. In lieu of these ASME Section XI administrative controls, Vistra OpCo will establish Owner-defined administrative controls as required by paragraph -1420(a) of Code Case N-752. Vistra OpCo will

Attachment TXX-24069 L-24-211 Page 19 of 24 utilize its existing work management processes for planning and documenting the performance of repair/replacement activities and supplement those process requirements as necessary to comply with Code Case N-752. These controls will ensure that repair/replacement activities on LSS items are performed in accordance with work instructions that have been appropriately planned, reviewed, and implemented. It should also be noted that the exemption of Repair/Replacement Plans as required by IWA-1400(k) and IWA-4150 is not unique to Code Case N-752.

Repair/Replacement Plans are already exempt by ASME Section XI for certain items and activities such as small items (IWA-4131) and rotation of items for testing or preventative maintenance (IWA-4132). Finally, the exemption of ASME Section XI programs and plans and the alternative use of Owner-defined administrative requirements on LSS items is consistent with the NRC's position on risk-informed programs as specified in 10 CFR 50.69(b)(1)(v).

16. As permitted by Code Case N-752, Vistra OpCo intends to implement the exemption on IWA-4000 applicable to repair/replacement activities. Article IWA-4000 of the ASME Section XI Code specifies administrative, technical, and programmatic requirements for performing repair/replacement activities on pressure-retaining items and their supports. As specified in IWA-411 0(b),

repair/replacement activities "include welding, brazing, defect removal, metal removal by thermal means, rerating, and removing, adding, and modifying items or systems. These requirements are applicable to procurement, design, fabrication, installation, examination, and pressure testing of items within the scope of this Division." In lieu of these IWA-4000 requirements, Vistra OpCo will perform repair/replacement activities on LSS items in accordance with an Owner-defined program that complies with paragraph -1420 of Code Case N-752. The Vistra OpCo program will utilize existing processes such as those applicable to procurement, design, re-rating, fabrication, installation, modifications, welding, defect removal, metal removal by thermal processes and supplement those process requirements as necessary to comply with Code Case N-752. Vistra OpCo believes this program will ensure, with reasonable confidence, that LSS items remain capable of performing their safety-related functions under design basis conditions. Finally, the exemption of IWA-4000 requirements and the alternative use of Owner-defined treatment requirements for LSS items is consistent with the NRC's position on risk-informed programs as specified in 10 CFR 50.69(b)(1)(v) and (d)(2).

17. As permitted by Code Case N-752, Vistra OpCo intends to implement the documentation exemptions on IWA-6211 (d), IWA-6211 (e), IWA-6211 (f),

IWA-6212, and IWA-6350. These ASME Section XI paragraphs address preparation and retention of various ASME Section XI records such as Form NIS-2, IWA-4160 verification of acceptability evaluations, IWA-4311 evaluations, Repair/Replacement Plans, and reconciliation documentation.

Attachment TXX-24069 L-24-211 Page 20 of 24 In lieu of these ASME Section XI forms and evaluations, the following repair/replacement activity records shall be retained in accordance with the Vistra OpCo Owner-defined program for performing repair/replacement activities on LSS items.

Repair/replacement activity documentation.

Evaluations of LSS items that do not comply with requirements of the applicable Construction Code, standard, specification, and/or design specification. See also paragraph 5.2.E.12.

Evaluations and documentation of design and configuration changes including material changes.

In addition to the above, Vistra OpCo will also revise applicable BVPS-1, BVPS-2, CPNPP-1, CPNPP-2, DBNPS, and PNPP licensing basis documents (e.g.,

Safety Analysis Report), as appropriate, to identify systems, subsystems, or individual items that have been categorized as LSS and address alternative treatment requirements. Changes to licensing basis documents will be performed in accordance with 10 CFR 50.59.

F.

Conclusion Code Case N-752 specifies requirements for performing risk-informed categorization and treatment for performing repair/replacement activities on Class 2 and 3 pressure-retaining items or associated supports. The Code Case N-752 categorization process provides a comprehensive methodology for determining the safety significance of items - HSS or LSS. This categorization process is technically identical to that approved by the NRC under relief request ANO2-R&R-004, Revision 1 (Reference 8.3). Repair/replacement activities performed on items determined to be HSS must continue to comply with the ASME Section XI Code. Repair/replacement activities performed on LSS items may comply with alternative treatment requirements that are defined by the Owner but must comply with all provisions of paragraph -1420 of Code Case N-752. Vistra OpCo's proposed treatment requirements, as described herein, meet these criteria, and provide reasonable confidence that LSS systems and items remain capable of performing their safety-related functions under design basis conditions. Finally, categorization and treatment requirements of Code Case N-752 applicable to repair/replacement activities are consistent with NRC requirements specified in 10 CFR 50.69.

Attachment TXX-24069 L-24-211 Page 21 of 24

6. Duration of Proposed Alternative

The duration of the proposed alternative is for the remainder of the current operating license for PNPP and current renewed operating licenses BVPS-1, BVPS-2, CPNPP-1, CPNPP-2, and DBNPS as shown below.

Docket License Unit Number License No.

Expires BVPS-1 50-334 DPR-66 01/29/2036 BVPS-2 50-412 NPF-73 05/27/2047 CPNPP-1 50-445 NPF-87 02/08/2050 CPNPP-2 50-446 NPF-89 02/02/2053 DBNPS 50-346 NPF-3 04/22/2037 PNPP 50-440 NPF-58 11/07/2026

7. Precedent 7.1 NRC Letter to Entergy Services, LLC, "Grand Gulf Nuclear Station, Unit 1; River Bend Station, Unit 1; and Waterford Steam Electric Station, Unit 3-RE:

Authorization of Proposed Alternative EN-RR-22-001 to Use ASME Code Case N-752, 'Risk-Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 Systems,Section XI, Division 1' (EPID L-2022-LLR-0054)," dated May 30, 2024 (Accession No. ML24151A236 [Package]).

7.2 NRC Letter to Duke Energy Carolinas, LLC, "Oconee Nuclear Station, Units 1, 2, and 3 - RE: Authorization of Alternative to Use RR-22-0174, 'Risk-Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 SystemsSection XI, Division 1' (EPID L-2022-LLR-0060)," dated December 13, 2023 (Accession No. ML23262A967).

7.3 NRC Letter to Entergy Operations, Inc., "Arkansas Nuclear One, Unit 2 -Approval of Request for Alternative ANO2-R&R-004, Revision 1, Request for Use Risk-Informed Safety Classification and Treatment for Repair/Replacement Activities in Class 2 and 3 Moderate and High Energy Systems (TAC No. MD5250)," dated April 22, 2009 (Accession No. ML090930246).

7.4 NRC Letter to Entergy Operations, Inc., Arkansas Nuclear One, Units 1 and 2 -

Request for Approval of Change to the Entergy Quality Assurance Program Manual (EPID L-2020-LLQ-0005), May 19, 2021 (Accession No. ML21132A279).

8. References 8.1 10 CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems, And Components for Nuclear Power Reactors," USNRC, 69 FR 68047, November 22, 2004.

Attachment TXX-24069 L-24-211 Page 22 of 24 8.2 U.S. Nuclear Regulatory Commission, Regulatory Guide 1.201, Revision 1, "Guidelines for Categorizing Structures, Systems, And Components in Nuclear Power Plants According to Their Safety Significance," dated May 2006.

8.3 Entergy Letter to NRC dated April 17, 2007, "Request for Alternative ANO2-R&R-004, Revision 1, Request to use Risk-Informed Safety Classification and Treatment for Repair/Replacement Activities in Class 2 and 3 Moderate Energy Systems," (ML071150108) as supplemented by letters dated August 6, 2007 (ML072220160), February 20, 2008 (Accession No. ML080520186), and January 12, 2009 (Accession No. ML090120620).

8.4 NRC Letter to Entergy Operations, Inc., "Arkansas Nuclear One, Unit 2 -Approval of Request for Alternative ANO2-R&R-004, Revision 1, Request for Use Risk-Informed Safety Classification and Treatment for Repair/Replacement Activities in Class 2 and 3 Moderate and High Energy Systems (TAC No. MD5250)," dated April 22, 2009 (Accession No. ML090930246).

8.5 EPRI Report TR-112657, Revision B-A, "Revised Risk-Informed Inspection Evaluation Procedure," EPRI, Palo Alto, CA: 1999.

8.6 NRC Letter to FirstEnergy Nuclear Operating Company, "Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2)- Risk-Informed lnservice Inspection (RI-ISi) Program (TAC Nos. MB5687 and MB5688)," dated April 9, 2004 (Accession No. ML040780805).

8.7 NRC Letter to FirstEnergy Nuclear Operating Company, "Beaver Valley Power Station, Unit Nos. 1 and 2 - Issuance of Amendments RE: Request for Adoption of TSTF-425, Revision 3, 'Relocate Surveillance Frequencies to Licensee-Control -

Risk Informed Technical Specification Task Force Initiative Sb' (TAC Nos. MF2942 and MF2943)," dated March 6, 2015 (Accession No. ML14322A461).

8.8 NRC Letter to FirstEnergy Nuclear Operating Company, "Beaver Valley Power Station, Unit Nos. 1 and 2 - Correction of Errors in Safety Evaluation Associated with License Amendment Nos. 301 and 190 (CAC Nos. MF3301 and MF3302; EPID L-2013-LLF-0001)," dated April 5, 2018 (Accession No. ML18065A403).

8.9 NRC Letter to Vistra Operations Company LLC, "Comanche Peak Nuclear Power Plant, Unit Nos. 1 and 2 - Issuance of Amendment Nos. 183 and 183 Regarding the Adoption of Technical Specifications Task Force Traveler TSTF-505, Revision 2 (EPID L-2021-LLA-0085)," dated August 22, 2022 (Accession No. ML22192A007).

8.10 NRC Letter to Vistra Operations Company LLC, "Comanche Peak Nuclear Power Plant, Unit Nos. 1 and 2 - Issuance of Amendment Nos. 187 and 187, Respectively RE: Adoption of 10 CFR 50.69, 'Risk-Informed Categorization and Treatment of

Attachment TXX-24069 L-24-211 Page 23 of 24 Structures, Systems and Components for Nuclear Power Reactors' (EPID L-2023-LLA-0057)," dated June 10, 2024 (Accession No. ML24120A363).

8.11 NRC Letter to Energy Harbor Nuclear Corp., "Davis-Besse Nuclear Power Station, Unit No. 1 - Issuance of Amendment No. 131 to Relocate Specific Surveillance Frequencies to a Licensee-Controlled Program (EPID L-2019-LLA-0252)," dated November 3, 2020 (Accession No. ML20280A827).

8.12 NRC Letter to FirstEnergy Nuclear Operating Company, "Davis-Besse Nuclear Power Station, Unit No. 1 - Issuance of Amendment No. 298 to Adopt National Fire Protection Association Standard 805 (CAC No. MF7190, EPID L-2015-LLF-0001),"

dated June 21, 2019 (Accession No. ML19100A306).

8.13 NRC Letter to FirstEnergy Nuclear Operating Company, "Perry Nuclear Power Plant, Unit No. 1 - Issuance of Amendment Concerning Adoption of TSTF-425, 'Relocate Surveillance Frequencies to Licensee Control' (CAC No. MF3720) (L-14-106),"

dated February 23, 2016 (Accession No. ML15307A349).

8.14 Nuclear Energy Institute NEI 05-04/07-12/12-06 Appendix X, "Close Out of Facts and Observations (F&Os)", February 2017.

8.15 ASME/ANS RA-Sb-2005, Addenda to ASME RA-S-2002, Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, American Society of Mechanical Engineers, New York, New York, December 2005.

8.16 Regulatory Guide 1.200, Revision 1, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,"

January 2007 (Accession No. ML070240001 ).

8.17 Nuclear Energy Institute NEI 05-04, "Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard," Revision 3, November 2009.

8.18 ASME/ANS RA-Sa-2009, Addenda to ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, February 2009.

8.19 Regulatory Guide 1.200, Revision 2, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,"

March 2009 (Accession No. ML090410014).

8.20 ASME/ANS RA-Sb-2013, Addenda to ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, September 2013.

Attachment TXX-24069 L-24-211 Page 24 of 24 8.21 NRC Letter to NEI, "U.S. Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix X to Guidance 05-04, 07-12, and 12-13, Close-Out of Facts and Observations (F&Os)," dated May 3, 2017 (Accession No. ML17079A427).